This invention relates to a control system for a nuclear reactor and more particularly to a control system for preventing unnecessary reactor trips.
The principles for the generation of power by a nuclear reactor have been well established and are well understood. Briefly, the reactor contains uranium or plutonium fuel elements in a core arrangement. Through the mechanisms of neutron absorption and nuclear fission of the uranium or plutonium large amounts of energy are released. This released energy manifests itself in the form of heat which is utilized to generate electricity. In the pressurized water reactor context, the heat is transferred to a primary coolant which continuously circulates through the core and carries the generated heat to a heat exchange boundary where a secondary coolant or working fluid is heated. Ordinarily the secondary coolant is water and is vaporized at the heat exchange boundary to produce steam. The steam is then circulated in a secondary system to a turbine for its ultimate use. The turbine is caused to turn at a predetermined rate and is connected to a generator for the ultimate transformation of the thermal energy into electrical energy.
All elements of this system are functionally interrelated. As an example, an increase in reactor power increases the rate of energy transfer to the primary coolant which in turn increases the rate of energy transfer to the secondary coolant causing more energy provided to the turbine for its ultimate transformation into electrical energy. Conversely, if less electrical energy is required, the energy requirements of a turbine diminish. The steam flow to the turbine is reduced and consequently the turbine utilizes less of the thermal energy being transferred to the secondary coolant and an energy backup results. Since less energy is being drawn from the steam supply system when the steam flow is reduced, both the temperature and pressure of the steam generator secondary side are caused to increase. The effect of this increase in secondary coolant temperature is reflected on the primary side of the heat exchanger since less energy can be transferred across the heat exchange boundary. Accordingly, both the primary coolant temperature and pressure increase. This trend continues until the reactor regulating system, which is programmed to keep the average temperature of the primary coolant on a specified program, returns the system to acceptable values by cutting back on the reactor's power by driving regulating rods into the core.
Since the regulating rods can only be slowly advanced into the core at a limited maximum speed, the reactor regulating system is unable to prevent a serious increase in primary and secondary pressures and temperatures if the magnitude and rate of energy backup described above exceeds certain values. Among other things, energy backup is dependent on the magnitude and rate of decrease in turbine load: called a turbine load rejection. Ordinarily, the reactor and steam supply systems are designed to be able to withstand a predetermined load rejection and rate of the load rejection. However, if the actual load rejection or rate of load rejection exceed these predetermined values, the reactor regulating system is unable to compensate rapidly enough for the energy backup and the temperatures and pressure of the primary system may increase uncontrollably. When this occurs, protective systems come into operation to trip the reactor and/or to open steam relief valves in order to avoid an overpressurization of primary and secondary systems. If the uncontrolled increase in pressure is not avoided by these measures, the safety pressure valves of either the primary or the secondary side are caused to lift. This is an undesirable occurrence since it may put the system out of operation until the seals of the safety valves have been remachined and reseated.
At this point it should be noted that the protective systems normally include both reactor protective systems and equipment protective systems. The reactor protective systems are designed to trip the reactor in the event certain limits are exceeded in the primary system such as too high of a primary pressure or too low of a DNBR. These are safety trips in the sense that they are required to insure only safe operation of the reactor. On the other hand, the equipment protective systems are designed to trip the reactor to prevent plant equipment damage. An example of such an equipment protective trip is the reactor trip on turbine trip, the rationale being that such an occurrence would normally result in the safety pressure valves eventually lifting and one of the reactor protective trip limits (usually high primary pressure) eventually being exceeded. In other words, the reactor is tripped at an early stage since such a reactor shutdown is eventually expected and since plant equipment damage (lifting of the safety pressure valves) can be prevented.
However, tripping the reactor unnecessarily is undesirable since it temporarily removes the nuclear power plant as a supplier and a time-consuming and therefore expensive reactor startup procedure must be followed before the reactor can be put back into operation as a power producer.
Systems have been devised to prevent overpressurization in the primary and secondary systems in the event of large or rapid load rejections. One such system is disclosed in copending application Ser. No. 347,260 filed Apr. 2, 1973 entitled "Steam Relief Valve Control System for a Nuclear Reactor" by Jose Marcelo Torres, assigned to the same assignee as the present invention. That application deals with the use of steam relief valves operable in response to rises in secondary system pressure to bypass excess steam to the condenser or to dump steam to the outside atmosphere. Therefore, by using a plurality of steam relief valves which are operable to prevent overpressurization in the primary and secondary systems, it is no longer absolutely necessary to trip the reactor everytime the turbine is tripped.